Office of Compensation Analysis and Support 

Technical Basis Document for Atomic Energy Operations at 

Blockson Chemical Company, Joliet, Illinois 

Subject Experts: Tom Tomes, Sam Glover 

Document Number: 

OCAS-TKBS-0002 

Effective Date: 11/21/2007 

Revision No.: 02 

Page 1 of 46 

Supersedes: 

11/21/2007 Revision No.: 01 Approval: Signature on File 

Date: 

James W. Neton, Associate Director for Science 

TABLE OF CONTENTS 

Section Page 

1.0 Introduction..........................................................................................................................4 

2.0 Site Description and Operational History............................................................................5 

2.1 Blockson Facilities Affected by Uranium Operations .............................................5 

2.2 Process Development ...............................................................................................6 

2.3 Uranium Recovery Operations ................................................................................7 

2.4 Radiological Data...................................................................................................11 

3.0 Internal Dose ......................................................................................................................12 

3.1 Radionuclide Source Term ....................................................................................13 

3.2 Uranium Intakes.....................................................................................................15 

3.2.1 Calcining....................................................................................................16 

3.2.2 Building 55.................................................................................................18 

3.2.3 Pilot Plant and Laboratory .........................................................................22 

3.3 Radon Exposures ...................................................................................................22 

4.0 External Dose .....................................................................................................................23 

4.1 External Dose from Existing Operations ...............................................................23 

4.2 External Dose from Uranium Extraction Operations.............................................23 

4.2.1 Source Term ...............................................................................................24 

4.2.2 Exposure from Drums of Uranium ............................................................26 

4.2.3 Exposure from Contaminated Surfaces .....................................................29 

4.2.4 Beta Dose ...................................................................................................31 

4.3 Occupational Medical Dose ...................................................................................32 

5.0 Dose from Residual Contamination...................................................................................33 

6.0 Dose Reconstruction Summary..........................................................................................39 

7.0 Attributions and Annotations .............................................................................................41
 

8.0 References..........................................................................................................................42 
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LIST OF TABLES 

Table Page
 

1 Building 55 Relative Radionuclide Concentrations...........................................................15 

2 Dust Concentrations in Air during Phosphate Industry Processes.....................................16 

3 Inhalation and Ingestion Rate for Calcining ......................................................................17 

4a Inhalation Rate for Building 55 .........................................................................................21 

4b Ingestion Rate for Building 55...........................................................................................22 

5 Radon Exposures ...............................................................................................................23 

6 Radioactivity Ratios for External Dose Modeling.............................................................26 

7 Dose Rates from Drums of Yellowcake ............................................................................27 

8 Annual Organ Dose from Exposure to Drums of Yellowcake ..........................................29 

9 Beta Dose to Skin...............................................................................................................32 

10 Annual Organ Dose from Medical X-rays.........................................................................32 

11 Annual Dose from Residual Contamination ......................................................................33 

12a Inhalation Intake Rate from Residual Contamination .......................................................35 

12b Ingestion Intake Rate from Residual Contamination .........................................................37 

13 Annual Radon Exposure from Residual Contamination....................................................38 

14 Dose Reconstruction Summary Table ...............................................................................40 

LIST OF FIGURES 

Figure Page
 

1 Photograph of the Joliet Plant of the Blockson Chemical Company.................................. 8
 

2 The Blockson Chemical Process for the Recovery of Uranium from Phosphoric Acid......9 

3 Distribution of Uranium Intakes ........................................................................................21 

4 Beta Dose Rate on the Surface of Yellowcake ..................................................................25 

5 Beta Dose Rate from Yellowcake Separated from Ore for More Than 100 Days as a 

Function of Distance from the Surface ..............................................................................25 

6 Effect of Density on Dose Rate at 30 cm from Drum of Yellowcake ...............................28 

7 Cumulative Dose Rate from Photons and Bremsstrahlung at 30 cm from a Drum
 

Containing U3O8 (density 4 g/cm3

)....................................................................................28 
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RECORD OF ISSUE/REVISIONS 

ISSUE 

AUTHORIZATION 

DATE 

EFFECTIVE 

DATE 

REV. 

NO. DESCRIPTION 

Draft 6/29/2006 00-A Document initiated to establish the 

technical basis for radiation dose 

reconstruction for former workers from the 

Blockson Chemical Company. 

7/31/2006 9/11/2006 00 Incorporates resolution of review 

comments. This supersedes ORAUTTKBS-0002. A PER is required for dose 

reconstructions performed with the 

superseded document. 

Draft 3/30/2007 01-A Rough draft for internal review. Added 

dose from additional locations on site and 

from additional radionuclides. Additional 

descriptions and editorial changes. 

Draft 4/11/2007 01-B Draft issued for review. 

Draft 6/14/2007 01-C Incorporated review comments. Revised 

internal and external dose modeling. 

Added additional site information and 

radiological data discussion. 

6/18/2007 6/20/2007 01 Revision 01. Changes results in an 

increase of dose, and a PER is required. 

11/21/2007 11/21/2007 02 Changed footnote 2 of Table 4a and 

footnote 3 of Table 12a to require 

consideration of Type M and Type S 

thorium in Building 55. Corrected an error 

in the liver dose in Table 8. Corrected 

errors in columns of Table 7 and resulting 

graph in Figure 6. No other changes were 

made. 
Effective Date: 

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1.0 Introduction 

Technical basis documents and site profile documents are not official determinations made by 

the National Institute for Occupational Safety and Health (NIOSH) but are rather general 

working documents that provide historic background information and guidance to assist in the 

preparation of dose reconstructions at particular sites or categories of sites. They will be revised 

in the event additional relevant information is obtained about the affected site(s). These 

documents may be used to assist NIOSH staff in the completion of the individual work required 

for each dose reconstruction. 

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic 

weapons employer facility” (AWE facility) or a “Department of Energy facility” as defined in 

the Energy Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 

U.S.C. § 7384I(5) and (12)]. EEOICPA, as amended, provides for employees who worked at an 

AWE facility during the contract period and/or during the residual contamination period to 

receive compensation if certain requirements are met. 

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when 

the AWE was processing or producing material that emitted radiation and was used in the 

production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination outside 

of the period in which weapons-related production occurred). For contract period employment, 

all occupationally-derived radiation exposures at covered facilities must be included in dose 

reconstructions. NIOSH does not consider the following exposures to be occupationally-derived: 

• radiation from naturally occurring radon present in conventional structures; and 

• radiation from diagnostic X-rays received in the treatment of work-related injuries. 

For residual contamination period employment, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) (i.e., radiation doses received from DOE/AEC-related work) must be 

included in dose reconstructions. Radiation dose received from DOE/AEC-related work 

includes: (1) radiation from radon consistent with NIOSH’s policies for including such radiation 

in the contract period; and, (2) medical screening X-rays, but not diagnostic X-rays for the 

treatment of work-related injuries. It should be noted that: (1) under subparagraph A of 

§ 7384n(c)(4), radiation associated with the Naval Propulsion Program is specifically excluded 

from the employee’s radiation dose; and, (2) under subparagraph B of this section, radiation from 

a source not covered by subparagraph A that cannot be reliably distinguished from radiation that 

is covered by subparagraph A is considered part of the employee’s radiation dose. This site 

profile covers only exposures resulting from nuclear weapons-related work. Exposures resulting 

from non-weapons related work, if applicable, will be covered elsewhere. 

Section 2.0 of this document describes the Blockson Chemical Company site and its history 

including some information about the radiological processes and source terms as well as the 

radiological controls and monitoring practices. Sections 3.0 and 4.0 discuss internal and external 

dose, respectively. 
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The following summary is to help provide consistency in dose reconstructions and to help assure 

that all components of dose are adequately addressed when doses are reconstructed. It also 

provides some information regarding the radiological processes and source terms, as well as 

information on the radiological controls and monitoring practices. 

This site profile provides specific information on historical practices and radiation exposures at 

the Blockson Chemical Company facility in Joliet, Illinois. 

2.0 Site Description and Operational History 

The Blockson Chemical Company manufactured wet-process phosphoric acid from Florida 

phosphate rock, which was subsequently used to manufacture other chemicals at the plant (Barr 

et al. 1955, Clegg and Foley 1958). In 1950-1951 the U. S. Atomic Energy Commission (AEC) 

approached several phosphate rock consumers about the possibility of recovering the uranium 

from the phosphate rock they processed. At the Blockson Chemical Company plant, the AEC 

was interested in the uranium that could be separated from the phosphoric acid. In early 1951 

the research staff at Blockson began an evaluation of the available research data and preliminary 

experimentation that the AEC made available to them. They determined that the only 

economically feasible approach applicable to the Blockson process would be to make the 

uranium recovery a by-product process from the existing operations (Stoltz, Jr. 1958). Building 

55 was constructed to house uranium recovery operations (Blockson 1951, DOE 2007). It 

operated through March 1962 (DOE 1983). The entire Blockson plant stopped production of 

chemicals in June 1991 (EPA 1993a). Building 55 was demolished in 1996 (Olin 2007). 

2.1 Blockson Facilities Affected by Uranium Operations 

The U. S. Department of Energy (DOE) lists the Blockson Chemical Company as an Atomic 

Weapons Employer from 1951 through 1962 with a Residual Radiation period from 1962 

through July 2006. The DOE covered facility description specifies Blockson operated Building 

55 to produce uranium for the AEC (DOE 2007). The listing further states that “This listing is 

also intended to cover the AEC-funded laboratory, pilot plant and oxidation process, which also 

occurred at Blockson, and was related to the work in Building 55.” The U. S. Department of 

Labor (DOL) sent NIOSH an interpretation of the areas they consider to be covered under 

EEOICPA, citing the current designation that was published in the Federal Register (FR) on 

August 23, 2004 (DOL 2007). The FR listed “Blockson Chemical Company (Building 55 and 

related activities)” as the covered facility (FR 2004). The DOL determined that the designation 

applies to “Building 55, the pilot plant and associated processes (including the oxidation process) 

and all the areas in which these processes where performed.” Consequently, NIOSH has 

conducted research for this revision of the site profile to identify changes made to operations on 

site to accommodate uranium extraction work. 

Blockson received phosphate rock primarily on barges. The rock was unloaded and put into 

silos. A conveyor transferred the rock to a large outdoor calciner (furnace), where it was heated 

to a high temperature. The purpose of the calcination was to break down organic material and to 

prepare the feed material for digestion. After calcination, samples were obtained and sent to the 
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laboratory for analysis. A former worker has said the sampling was to test for fineness to 

determine if the contents of the furnace were acceptable for transfer to Building 40 or if 

additional calcination was needed (OCAS 2007a, OCAS 2007b, Stoltz, Jr. 1958). Calcination 

was already being performed at Blockson prior to AEC work; however, changes to calcining 

procedures were made when Blockson began uranium extraction work. Calcination would leave 

uranium and some other elements in a reduced state. Oxidizing conditions were desired to 

improve uranium recovery in subsequent processing steps. In previous years Blockson had done 

extensive research on calcination. When uranium extraction started, they pulled out the old 

research and modified the calcination procedure to help maintain uranium in an oxidized state 

(Stoltz, Jr. 1958). Therefore, calcining operations at Blockson can be considered a related 

activity and were affected by the AEC contract work, and for the purpose of dose reconstruction 

are considered in this site profile. 

A conveyor transferred the calcined phosphate feed to the acid plant, Building 40 (once known 

as Building 25) (OCAS 2007b). Blockson modified their existing operations by adding an 

oxidizer to the phosphoric acid to increase uranium recovery in subsequent steps. Following the 

flow of raw material into the plant, this is the second operational change known to have occurred 

as a result of AEC work. Several oxidizers were tested and chlorine was chosen to be added to 

the phosphoric acid (Stoltz, Jr. 1958). Monosodium phosphate was produced from the acid. For 

the AEC work, Blockson added a filter, tank and other equipment to increase capacity so all 

Blockson phosphate could be diverted through the uranium recovery process (Blockson 1951). 

Former workers have indicated that the feed to Building 55 came from Building 40 (OCAS 

2007a). Therefore, since chlorination was performed on the acid and the building equipment was 

altered somewhat, Building 40 is considered in this profile for the purpose of dose 

reconstruction. 

Laboratory work was performed to develop the Blockson process for uranium extraction. All of 

the AEC funded laboratory work on process development leading up to construction and 

operation of Building 55 was documented in a report (Blockson 1953a). Specifics on the 

laboratory facility are not available, although the process work is described in detail. The work 

consisted of various procedures and reagents used to precipitate uranium from the existing 

processes at Blockson. The source (phosphate rock and derived products) contained the same 

radioactive constituents as the product processed in the areas described above and eventually in 

Building 55. During the operational period Blockson also analyzed samples collected from the 

pans used to dry the uranium concentrate in Building 55 (Blockson 1953b). 

A pilot plant was constructed concurrent with the laboratory developmental work. Pilot plant 

work and results are also documented by Blockson (1953a). Details are provided that include 

parameters used for process steps, dates of operation, and uranium analytical results for the 

various procedures and steps being studied. The laboratory work and pilot plant operations have 

been considered in this site profile for the purpose of dose reconstruction. 

2.2 Process Development 

On March 6, 1951, the AEC entered into letter contract number AT(49-1)-606 with Blockson 

Chemical Company to develop a process to extract uranium from wet phosphoric acid (DOE 
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1983, DOE 2007). Laboratory studies began at Blockson at about that time (Stoltz, Jr. 1958), 

and due to the urgency of the program, pilot plant construction was begun simultaneously with 

the start of laboratory work (Blockson 1953a). Documentation indicates that pilot plant runs 

lasted about three weeks, ran 24 hours a day at 25% of expected production capacity, and that 

several runs were performed: April 29, 1951, to June 21, 1951; July 23, 1951, to August 10, 

1951; and November 25, 1951, to January 6, 1952 (Blockson 1953a). Various uranium recovery 

methods were investigated and tested through pilot plant runs, and by July 1951 Blockson had 

determined the most effective process (Blockson 1951). The building in which Blockson 

constructed the pilot plant is not known. 

Blockson sent a letter to the AEC on July 31, 1951 (Blockson 1951) in which the uranium 

recovery process was summarized. The letter included best estimates of production, fixed capital 

requirements, manufacturing costs, and general contract conditions which would be acceptable to 

Blockson Chemical Company. This letter states that all process equipment could be housed in 

the new building (Building 55) with the exception of an 8x12 filter, a storage tank, and pumps. 

The new filter was necessary to enable all Blockson’s liquors to be processed through the 

monosodium phosphate step, which was the liquor to be diverted to Building 55. A monosodium 

phosphate liquor storage tank was added to provide storage capacity for monosodium liquors 

from which the uranium had been removed. These two items, with associated pumps and other 

accessories, were to be located in their existing plant buildings where those operations were 

already being performed. 

The letter contract was later replaced by contract number AT(49-1)-611 on October 18, 1951 

(AEC 1951). Under the contract Blockson began construction of Building 55 at its own expense 

to house uranium recovery equipment at their plant in Joliet, Illinois. Laboratory work and pilot 

studies continued during construction of Building 55 in efforts to further improve uranium 

recovery processes, but no significant changes in the process were made. 

2.3 Uranium Recovery Operations 

The recovery plant was put into operation on August 15, 1952. The process was patented and 

the patent, USP 2743156, was assigned to the AEC (Stoltz, Jr. 1958). Building 55 was a onestory, 100-by-175-foot building built specifically to house the uranium recovery process (AEC 

1951, DOE 1983). A photograph of Building 55 is shown in Figure 1. 
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Figure 1: Photograph of the Joliet Plant of the Blockson Chemical Company. 

[Reproduced from Barr et al. 1955]. 

Blockson Chemical Company produced technical phosphates rather than fertilizers from wet 

phosphoric acid (Wilkinson 1976). In the Blockson process, the phosphate rock was calcined, 

pulverized, and then digested with sulfuric acid resulting in phosphogypsum and phosphoric 

acid. Blockson modified its calcining procedure to decrease losses of uranium to the 

phosphogypsum. The phosphoric acid partitioned about 85% of the uranium while the 

phosphogypsum partitioned most of the calcium and radium (Blockson 1951, Stoltz, Jr. 1958). 

The phosphoric acid was partially neutralized with soda ash to form monosodium phosphate. 

The acid contained uranium in both the uranous (4+) and uranyl (6+) state. During 

neutralization of the acid uranous ions tended to partially precipitate when the pH was increased. 

To increase the recovery of uranium in the monosodium phosphate liquor, Blockson added a 

chlorination step to the phosphoric acid. Calcium, aluminum and iron salts were formed during 

conversion, were filtered off, and about 90% of the uranium was recovered in the liquor. The 

monosodium phosphate liquor was pumped to the uranium recovery building (Building 55) 

(Blockson 1951, Blockson 1953a, Stoltz, Jr. 1958). 

The uranium by-product was precipitated from the monosodium phosphate stream that was 

diverted to Building 55. In Building 55 monosodium phosphate was “heated and clear pressed to 

remove any fine suspended solids” (Blockson 1951). Sodium hydrosulfite (Na2S2O4) was added 

to precipitate the uranium. A filter aid was also added to help remove the precipitant from 

solution. The filtrate was returned to the phosphate-processing plant and the precipitate, 

containing about 5% U3O8, was further processed to up-grade the product. The precipitate was 

slurried in a weak sulfuric acid solution with a small amount of chlorine which resulted in a 

gradual dissolution of the uranium. The filter aid was then filtered off and the filtrate partially 

neutralized to precipitate the uranium. The uranium was then filtered as uranous phosphate 

containing 40% to 60% U3O8. The product was dried and packaged for shipping (Blockson 

1951, Stoltz, Jr. 1958, Clegg and Foley 1958). The material precipitated in Building 55 has been 
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reported to be a fine powdered sodium uranium compound (Wimpfen 2002). Other reports have 

indicated that the precipitated compound was sodium di-uranate (Cope and Sinclair 1952) and a 

uranous phosphate (Clegg and Foley 1958). The uranium content of the phosphate rock 

consumed in these processes was reported to be as high as 0.014% U3O8 (Stoltz, Jr. 1958). 

Figure 2 shows a schematic flowchart of the Blockson uranium recovery process. 

Figure 2 The Blockson Chemical Process for the Recovery of Uranium from Phosphoric 

Acid [reproduced from Clegg and Foley 1958]. 

Accounts from former Building 55 workers (OCAS 2007a) indicate that the precipitated and 

filtered uranium concentrate in Building 55 was placed in pans and pictures and descriptions 

indicate the pans nominally measured two and one half feet by two and one half feet and a 

couple inches thick. The pans were placed in a drier overnight, then emptied into a drum. This 

was a manual operation that was normally performed on the day shift when there were additional 

workers. The former workers are in general agreement that there were about two to six full time 

Building 55 workers per shift, with the night shift being minimally staffed. Blockson’s preoperational labor estimates submitted to the AEC allowed for two operators and one chemist per 

shift and allowed for two additional men on day shift. Additionally, they estimated an average of 

two full time mechanics and budgeted for a clerk, a developmental chemist, and a foreman 

(Blockson 1951). The former workers are in agreement that a number of maintenance people 

were in and out of Building 55 performing various jobs within their area of expertise (OCAS 

2007a). 
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A pre-operational letter from Blockson to the AEC in July 1951 provided estimates of 

construction and operating costs based on Blockson processing 6,000 tons of phosphate rock per 

week (Blockson 1951). Taking recovery into consideration they estimated producing 50,000 

pounds of U3O8 per year. AEC reported in 1955 that U3O8 production from Blockson was 3,758 

pounds in August; 3,407 pounds in September; 5,908 pounds in October; 4,093 pounds in 

November; and 2,937 pounds in December of 1955 (AEC 1955a, 1955b, 1955c, 1955d, 1955e). 

This indicates an average production rate of over 4,000 pounds per month in the latter part of 

1955. The December report included information that Blockson had produced a total of 121,400 

pounds in the 40 months of Blockson operations to that point, an average of about 3,035 pounds 

of U3O8 produced per month through 1955. Information from a former AEC official (OCAS 

2007c) indicates that from September 1952 through June 1960, a total of 118.3 tons U3O8 had 

been produced for the AEC. This equates to a little less than 3,000 pounds of U3O8 per month 

through June 1960. This is consistent with a 3,000 pound monthly production estimate from 

1958 found in a letter to the Feed Material Production Center, which was receiving the Blockson 

product at that time and sampling it to determine delivery quantities for payment purposes (AEC 

1958a). The total amount of uranium produced from 1960 through the end of operations in 1962 

is not precisely known. Per a contract amendment in 1958 production was limited to 50,000 

pounds of U3O8 per year (US DOE 1985). 

There is contradictory information in some literature regarding the total uranium production 

through 1955. A Department of Energy Formerly Utilized Sites Remedial Action Program 

(FUSRAP) Report (DOE 1985) indicates that 1.22 million pounds of uranium concentrates had 

been produced by the end of calendar year 1955. Some newspaper accounts have reported that 

Blockson produced about 2 million pounds of uranium throughout the AEC contract period. The 

amount reported in the FUSRAP report is apparently an overestimate by a factor of 10 based on 

the total production through 1955 as described above from detailed AEC production reports. For 

example, in order to produce 1.22 million pounds of U3O8 in that timeframe, Blockson would 

have had to double the capacity of the entire facility, used phosphate ores with the highest known 

uranium content (0.03%), and extract 100% of the uranium throughout each of its processes. In 

reality, uranium content of Florida phosphate ore averages 0.011% and maximum recovery for 

uranium at Blockson was estimated to be 73%, which results from recoveries of 85% for rock 

digestion, 90% for monosodium phosphate precipitation, and 95% for U3O8 precipitation and 

upgrading steps (Blockson 1951). Those values, along with a throughput of 6,000 tons per week 

were used to estimate the production at 50,000 pounds of U3O8 per year. In early 1953 Blockson 

reported up to that time they had actually achieved a recovery of 60% - 70% (Blockson 1953a). 

The 1955 AEC reports and the quantities produced through June 1960 (OCAS 2007c) indicate 

that, on average, Blockson produced less than 35,000 pounds per year, which is less than the preproduction estimate of 50,000 pounds per year. Documentation also indicates that on at least one 

occasion uranium recovery at Blockson was less than planned. Blockson attributed these lower 

recoveries to differences in the phosphate rock feed received by the plant. 

In 1955, Blockson was sold to the Olin Mathieson Chemical Corporation (later renamed Olin 

Corporation), who assumed the liabilities and obligations of Blockson. As stated in contract 

number AT(49-1)-611 Amendment 1, Olin continued the uranium recovery program under 
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contract with the AEC. In March 1962, the uranium recovery work was discontinued with the 

expiration of the contract (DOE 1983). 

According to the contract signed in October of 1951, Blockson, and later Olin Mathieson, was 

responsible for the health and safety of the employees at the site and for conforming to AEC 

health and safety regulations and requirements. In Amendment 3, effective January 1, 1958, this 

statement was deleted. 

According to former Blockson workers the AEC maintained a guard and controlled access to 

Building 55. The workers also indicated that the security was eased in the late 1950s (OCAS 

2007a). A former worker provided a copy of an agreement effective January 15, 1957, between 

Blockson and the International Chemical Worker’s Union Local No. 4. Article 33 of that 

agreement discusses the elimination of security provisions for Department 55 and provisions for 

pay and movement of union personnel into and out of that department. Relaxing of security 

requirements was also indicated in a May 1955 AEC memorandum which said a new security 

classification was being written for uranium phosphate operations that “may completely 

declassify everything except overall production data” (Robinson 1955). It is not known exactly 

when AEC security requirements were eased at Blockson, but the statements from the former 

workers are supported by the AEC’s plan to declassify some of the operations. 

2.4 Radiological Data 

The only radiological data from the AEC operational period that is known to exist are bioassay 

results for uranium from 1954 through 1958. The bioassay results are discussed in Section 3. 

Personnel with the FUSRAP Program conducted record searches for information regarding the 

uranium recovery activities at Blockson. No records of health and safety inspections by the AEC 

were found as a result of their search, although there was evidence of periodic visits by AEC 

personnel to review and audit process operations (DOE 1985). Records of bioassay monitoring 

have been found during data capture efforts for the EEOICPA program, but there is no indication 

from records searches that external dosimetry was utilized at Blockson. The FUSRAP program 

also conducted extensive monitoring of Building 55 in 1978. That data is discussed in other 

sections of this document. 

The Olin Corporation has provided radiological data from surveys conducted in 1983 and 1996 

(Olin 2007). The 1983 data consists of total dust measurements, airborne alpha radioactivity 

measurements, and radon working level (WL) measurements. The 1983 measurements were 

from several areas but none from Building 55. NIOSH has also obtained copies of 1993 radon 

flux measurements of the phosphogypsum stacks that were performed for Olin (Olin 1993). The 

radon estimates in this document are based on a favorable evaluation of WL measurements from 

phosphate plants with similar operations. 

There are two sets of radiological data from 1996. One set of 1996 data is from RSSI who wrote 

a Radiological Safety Manual for demolition of other buildings unrelated to Building 55. The 

other 1996 data is from a survey conducted in June by Radiation Safety Associates. The survey 

was performed on Building 55, which was partially dismantled at the time. Surveys were of 
 

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various building and equipment components and surrounding soils. Isotopic data from those 

surveys are discussed in the internal dose section. The contamination levels from that survey are 

consistent with the values and assumptions used in estimated dose from residual contamination 

based on the 1978 survey discussed in the section on residual contamination. 

3.0 Internal Dose 

Because phosphate rock contains Naturally Occurring Radioactive Material (NORM), work 

involving phosphate rock potentially exposes workers to radioactivity. This includes potential 

exposure to airborne uranium and its progeny, and other radiological constituents that may be 

present. Handling phosphate rock prior to chemical processing could have produced an airborne 

concentration of uranium dust that could contain all of its progeny in equilibrium. Internal dose 

has been evaluated based on two likely exposure scenarios: dose from uranium extraction work, 

and dose from other work. Radon exposures are also considered. 

The primary uranium isotopes in the phosphate rock are U-238, U-234, U-235 and associated 

progeny. Dosimetrically significant progeny include Th-230, Ra-226, Rn-222 (radon) and 

associated progeny. Trace amount of natural thorium (and associated progeny) are also present 

in phosphate rock (FIPR 1995). 

Uranium extraction work potentially exposed workers to uranium that was significantly 

concentrated. Uranium concentration in the precipitated final dried product produced for the 

AEC was about four thousand times higher than the concentration in the phosphate rock received 

by Blockson. Reports of the average uranium concentration in the phosphate rock used by 

Blockson ranged from 0.011% to 0.014%, while the uranium concentration in the final 

precipitation in Building 55 was reported to be 40% to 60% (Blockon 1951, Stoltz, Jr. 1958). 

Other steps in the extraction process produced a much lower concentration of uranium. The 

uranium extraction work occurred in the laboratory (for research and development, and for 

analyses), in the pilot plant (for production scale testing of laboratory results), and finally full 

scale operations in Building 55. Building 55 operated at a much larger capacity and potentially 

exposed workers to much larger quantity of uncontained radioactive material. Therefore, 

Building 55 intakes are used to bound intakes from uranium extraction work. Other 

radionuclides in the uranium extraction work that could have significant impact on internal dose 

are also considered. 

The second internal dose exposure scenario considered in this evaluation is the maximum dose 

likely to have been received outside of uranium extraction operations. The most significant 

radioactive elements at Blockson that did not follow uranium through the recovery process are 

radium and polonium. Most of the radium and polonium was precipitated and filtered off with 

the calcium during phosphoric acid production in Building 40, which was slurried and pumped to 

the phosphogypsum piles. Ra-226 is not as significant for internal dose as some of the other 

uranium progeny present in the rock, such as Th-230. Dose from Th-230 (and other thorium 

isotopes and progeny that are assumed to report to the phosphoric acid) is accounted for by 

assuming all of the thorium follows the uranium to Building 55, and is concentrated to produce a 

more significant internal dose hazard. Handling the radium bearing phosphogypsum was a wet 

operation and would produce a much less significant airborne concentration than handling dry 
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product, although some dry material would be expected. Significant amounts of dry material 

would not have accumulated in the plant over time due to Blockson requiring each shift to hose 

down the work areas (OCAS 2007a). Bounding internal doses from non-uranium extraction 

work is based on work that was likely to produce the highest airborne radioactivity 

concentrations, which was handling dried phosphate rock prior to chemical processing that 

removed some of the radiological components. 

3.1 Radionuclide Source Term 

Blockson Chemical processed phosphate rock that was mined in Florida. This rock then 

underwent a number of processing steps, including ore receiving and storage, calcining, 

crushing, digestion to produce phosphoric acid that was used to produce other compounds of 

phosphorous, as well as the by-product uranium (Stoltz, Jr. 1958). Hull and Burnett (1996) 

reported that U-238 and the various progeny are in approximate equilibrium in phosphate rock. 

Steps that involve only physical processing of the materials, e.g. crushing and storage, do not 

typically alter the relative concentrations of uranium, thorium, radium, or other radionuclides in 

the raw materials. However, chemical processing, e.g., phosphoric acid production, may alter 

this equilibrium and the various radioisotopes will fractionate according to chemical properties 

(FIPR 1995). Consideration must be given to the effect chemical processing has on the ratios of 

the various radionuclides present in the phosphate rock. 

It can be assumed that the various isotopes of a particular element will react similarly to 

chemical processing if they are in the same physical and chemical form. For example, if all of 

the thorium isotopes are present in the phosphate rock in a similar matrix distributed throughout 

the rock, then it can be assumed that all isotopes of thorium will remain in the same relative 

concentration as the rock is chemically processed. This may not, however, be a valid assumption 

if some thorium isotopes are bound in a different matrix in the phosphate rock. FIPR 1995 

reported in their study that unreacted minerals in the matrix account for an estimated 5-10% 

uncertainty in the fraction of the radionuclides reporting to the phosphogypsum. 

The distribution of specific uranium and thorium decay chain radionuclides within phosphate 

source materials and within the various products and waste streams produced by the phosphate 

ore processing industry has been the subject of numerous studies. While distribution of 

radionuclides is in some respects are a function of the specific process involved, some 

generalizations may be made specific to the process employed at Blockson: 

•	 Radiological equilibrium in the uranium chain appears to be maintained in rock that has 

not been chemically processed (Roessler 1979, FIPR 1995). 

•	 Ra-226 and Po-210 are retained in the phosphogypsum, i.e., do not enter the phosphoric 

acid stream to any significant degree (Roessler 1984 p7; Guimond 1975 p15; FIPR 1995 

p. 1-16). 

•	 Uranium and thorium tends to favor the phosphoric acid phase (Roessler 1984; Guimond 

1975, FIPR 1995). 

•	 Since Th-230 is present in the matrix with U-238, it is expected to go into solution along 

with the uranium when leached in sulfuric acid. Th-232, if occupying a different matrix 

in the rock, may not be as readily dissolved in sulfuric acid (Coppinger 1959 p. 20). 
 

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•	 Pb-210 is reported by some authors (FIPR 1995 p 1-16) as being retained in the 

phosphogypsum and by others as reporting to the phosphoric acid (Roessler 1984 p7). 

When Olin dismantled Building 55 in 1996 they stopped work and performed radiological 

surveys and took four media samples of debris in Building 55. They also took several soil 

samples around Building 55. These samples were obtained when the building had equipment, 

concrete, roof, and other structures partially demolished. The samples were analyzed by gamma 

spectroscopy and results were reported for a number of radionuclides in the U-238, U-234, and 

Th-232 (natural thorium) chain. Soil sample results were typically low. The four samples from 

inside Building 55 were taken of spots that were identified as contaminated based on scan 

results. These data are insufficient to form a quantitative conclusion on the ratios of the various 

radionuclides that were present, but the data indicates elevated uranium with some natural 

thorium, but at a much lower concentration. Ra-226 was not reported as being detected in those 

samples, nor was Pb-210, but low concentrations of Pb-214 and Bi-214 were indicated. Ra-226 

and Pb-210 were reported in low concentrations in some of the soil samples. One Building 55 

sample from the radiological characterization of the Blockson site was described as a ‘yellow 

powder’ had detectable quantities of Th-230 (Olin 2007). 

This data indicates that uranium isotopes were present with lesser amounts of natural thorium. 

Radium and lead were not reported from inside Building 55 (Pb-212 was reported on one sample 

and is expected due to ingrowth from natural thorium). However, this data is not considered 

sufficient to quantify the amounts of the various radionuclides for which workers could have 

been exposed during and after uranium extraction operations. 

The Blockson process was specifically engineered to maximize carryover of uranium into the 

phosphoric acid phase. Other elements would be present at various stages according to their 

chemical properties. There are uncertainties with chemical recoveries and potential losses of 

some elements in some of the chemical steps. In lieu of this uncertainty, an exposure model is 

provided that makes assumptions that result in limiting doses from the various radionuclides that 

are present in the feed material. 

Assumptions for isotopic ratios in the phosphoric acid stream and monosodium phosphate 

pumped to Building 55: 

1. 85% of U reports to phosphoric acid (Blockson 1951, Blockson 1953, Stoltz, Jr. 1958). 

2. 4% of Ra-226 reports to acid (Hull and Burnett 1996). Various publications have reported 

lesser or more percentage goes to the acid. Roessler et al., 1979, reported 1% reports to the 

acid; other studies have reported virtually no radium reports to the acid, while some have 

reported more. This document assumes 4% based on Hull and Burnett 1996. No Ra-226 

was detected in the surveys done in the later years, although Th-230 and natural thorium was 

detected as discussed above. 

3. Thorium reports to the acid in same proportion as uranium. Several references indicate the 

thorium is likely to be somewhat lower than uranium. The assumption of equal recovery of 

thorium to uranium in the acid results in a higher source term for internal and external dose 
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modeling. If there were more thorium losses to the phosphogypsum stream the doses would 

be lower. 

4. U-238:Th-232 radioactivity ratio in Blockson's rock was 30:1. This is considered a bounding 

ratio to allow for natural thorium and progeny based on reported U-238 and Th-232 

concentrations in phosphate rock (ORAU 2006). Th-232 progeny are assumed to be in 

equilibrium. Although most of the Ra-228 would have been separated and removed with the 

phosphogypsum, it is assumed to be in equilibrium with Th-232 for dose modeling to allow 

for ingrowth over the operational and residual contamination period. 

5. Pb-210 is assumed to report to the acid at 100%. The various references cite data indicating 

lead reports to the phosphogypsum, while others report high percentages reporting to the 

phosphoric acid. For modeling purposes, Bi-210 and Po-210 are assumed to be equal to Pb210 for the purpose of bounding exposures and intakes from ingrowth over the operational 

and residual period. 

5. All isotopes reporting to the acid are carried through to the drum of dried uranium 

concentrate (highest potential source for internal dose and source for external dose model) in 

the same relative concentration as U-238. 

These assumptions result in the following ratios for the liquor sent to Building 55: 

Table 1: Building 55 Relative Radionuclide Concentrations. 

Radionuclide 

Relative 

Ratio1 Notes Normalized to U-2381 

U-238 85 Progeny in equilibrium 

through Th-230 

1 

U-235 3.87 Progeny in equilibrium 0.0455 

Ra-226 4 Progeny in equilibrium 0.047 

Pb-210 85 Equal to U-238 1 

Bi-210 85 Equal to U-238 1 

Po-210 85 Equal to U-238 1 

Th-232 2.8 Progeny in equilibrium 0.033 

1. Ratios given are for progeny without consideration of branching ratios, where applicable. 

3.2 Uranium Intakes 

Intakes of radioactive material have been evaluated for uranium extraction operations and for 

other work at the Blockson Plant, some of which is considered to be associated with uranium 

recovery work, although not directly involved with uranium extraction. The calciner has been 

chosen to provide bounding intakes for non uranium extraction work as discussed below; it can 

be considered an associated activity because Blockson made a change to the procedure to 

improve uranium recovery in other processes. 
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3.2.1 Calcining 

The 1983 survey (Olin 2007, p.7) provided results of 12 general area and personnel breathing 

zone total dust measurements at several locations in the Blockson plant, none of which are from 

Building 55. The maximum reported total dust result from the 1983 study was 6.37 mg/m3

 and 

was associated with TSP (trisodium phosphate) operations. Gross alpha measurements of the air 

samples were all less than the sensitivity of the equipment except for a filter platform in Building 

40. All of these results are less than the intakes provided below based on a more extensive study 

done at another wet process phosphate plant. Therefore, the maximum result from the study 

described below is used to bound intakes at Blockson from work outside uranium extraction 

operations. 

The US EPA performed a thorough study of dust loading and radionuclide concentrations in air 

throughout phosphate ore processing at an Idaho Phosphate facility (EPA 1978). The facility 

utilized the wet-process method to process phosphate rock, the same method used by Blockson. 

Various air samples were taken at locations throughout the plant. The samples were analyzed for 

total dust loading and airborne radioactivity concentrations were reported. The atmospheric dust 

concentrations were not provided, but have been calculated from data. Specifically, the total 

reported mass of dust on the filter paper in mg was divided by the volume of air sampled to 

derive the atmospheric dust loading in units of mg/m3, which are listed in Table 2. The highest 

dust concentration in that study was from Calciner #3, which was 50.4 mg/m3

, indicative of an 

operation with likely visible dust. Information in the report indicates that it was located in the 

Calciner Building, although no other information on the calciner was provided in the report. 

Samples were also taken from several locations in the phosphoric acid building. 

Table 2: Dust Concentrations in Air during Phosphate Industry Processes 

Material Dust Loading (mg/m3) 

Grinder mill 6.29 

Ore unloading, storage 5.43 

Calciner control room 1.73 

Calciner 50.4 

Phos. Acid digester 15.3 

Outside control room 6.5 

Continuous filter 2.72 

Control room 1.5 

TSP storage 1.37 

TSP dryer 33.9 

Acidulation TSP disch. 2.63 

200 Ammophos plant 18.6 

100 Ammophos plant 8.68 

Library 0.92 

100 plant storage 8.61 

200 plant storage 4.12 

Blockson operated a calciner to break down organic matter and to roast the rock into a form that 

optimized recovery of uranium and for other processing purposes unrelated to uranium extraction 

work. Their calciner was a large outdoor furnace that utilized a dust collector (OCAS 2007b). 
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The dust concentrations at Blockson’s calciner would be expected to be lower than the one listed 

in Table 1 because of dilution with open air. The use of a dust collector at Blockson also 

indicates that some effort was made to control dust. Although it could be expected that some 

localized areas may have had high concentrations for short durations, such as when sampling the 

calcined rock, average dust concentrations would be expected to be lower than that reported for 

the operations at the phosphate plant discussed above. Therefore, these dust concentrations are 

being used to determine bounding dust concentrations for operations at Blockson outside of 

Building 55. 

Intake quantities for uranium and thorium nuclides (and associated progeny) were calculated 

based on continuous exposure to an airborne dust loading of 50.4 mg/m3 . Bounding intakes from 

uranium and thorium (and associated progeny) were calculated based on a breathing rate of 1.2 

m3

/hr and being exposed to that high level of dust for 2000 hours per year. For these 

calculations the uranium mass was assumed to be all U-238, as it represents over 99% of the 

uranium mass in natural uranium. U-234, Th-230, Ra-226, and other nuclides in the U-238 chain 

are assumed to be in equilibrium with U-238 during phosphate rock crushing and calcining 

operations. U-235 was not included because the method of calculating uranium activity allowed 

for the small intake contribution from U-235. Additional parameters used include an assumed 

0.014% uranium content in phosphate rock (Stoltz, Jr. 1958). Intake values have been 

normalized to calendar days. Ingestion intakes were also calculated using the inhalation intake 

quantities and applying the methodology in OCAS-TIB-0009 (Estimation of Ingestion Intakes). 

Polonium has a boiling point of 962oC, or 1764oF (Webelements 2007). Calcining temperatures 

in other phosphate facilities are known to have increased airborne particulate concentrations of 

Po-210 due to the high temperature used in the furnace. Radiological data from the Idaho 

phosphate facility calciner mentioned above indicated that the Po-210 concentration in air at 

the calciner was 10 times higher than the U-238 concentration (EPA 1978). Elevated Po-210 

concentrations from phosphate facility calcination has also been reported by Boothe (1977). 

Therefore, the intakes of Po-210 from calcining are estimated to be 10 times higher than intakes 

of U-238. Intakes of concern for internal dose are listed in Table 3. 

Table 3: Inhalation and Ingestion Rate for Calcining.1,2

 Inhalation Ingestion 

Radionuclides pCi/day pCi/day 

U-238, Th-230, U-234, 

Ra-226, Pb-210 

16 0.47 

Po-210 160 4.7 

Th-231, Pa-231, Ac-2273

 0.73 0.021 

Th-232, Ra-228, Th-228 0.52 0.016 

1. Intake rates have been normalized to calendar days. 

2. Both inhalation and ingestion should be assigned for calcining 

operations. 

3. U-235 is allowed for in the U-238 and U-234 values. Values given are 

for radionuclides in the U-235 chain. 

Except for Po, Pb, and Ra, the solubility class in the calcined phosphate rock may be Type M or 

S and should be selected to be claimant-favorable. Po is either Type F or M, Pb is assumed to be 
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Type F, and Ra is assumed to be Type M (ICRP 1994b). The inhalation solubility class and 

ingestion absorption values should be selected based on values that result in the highest organ 

dose. 

3.2.2 Building 55 

There are no air sampling results available from Blockson to characterize airborne radioactivity 

concentrations, however, air sampling results from mills with some common activities have been 

published. The greatest potential for internal uranium exposure in the Blockson uranium 

recovery process is most likely associated with handling dried uranium compounds in the 

packaging areas. Here the uranium concentrate (yellowcake) was dried and barreled for 

shipping, resulting in a potentially dusty operation (Eidson and Damon 1984, US NRC 2002a, 

Wimpfen 2002). 

A study was done (Eidson and Damon 1984) of uranium aerosols generated during yellowcake 

packaging operations at four uranium mills. The precipitated yellowcake is dewatered in a filter 

process, and then dried in an oven to produce powdered yellowcake that is placed in 55-gallon 

drums. Blockson also used a dryer to dewater the yellowcake prior to packaging in drums 

(Clegg and Foley 1958, Blockson 1951, OCAS 2007a). 

Eidson and Damon’s study described a sequence of steps common to all four uranium ore 

processing mills: 

1.	 No activity. This is when the plant is shut down for maintenance or when all available 

yellowcake has been barreled. Worker exposure to airborne yellowcake is minimal at 

this step. 

2.	 Barrel loading. This occurs when a barrel is placed under a hopper containing the dried 

yellowcake. The yellowcake is allowed to fall into the barrel. The amount of time 

workers spend in this area depends on the volume of the yellowcake in the hopper. 

3.	 Barrel uncovering. This step occurs when a filled barrel is removed from beneath the 

hopper. In some cases, the barrel may be vibrated to compact the yellowcake before 

removing the barrel from beneath the hopper. (It is not known if the barrels at Blockson 

were vibrated.) 

4.	 Powder sampling. This occurs when a worker takes a sample of yellowcake for 

laboratory analysis. At Blockson this was done prior to the pans of yellowcake being 

dumped into the barrels. 

5.	 Lid sealing. This occurs when a worker places a lid on the barrel and seals it. 

6.	 Other activities. This step includes maintenance and cleaning of the area with water 

hoses. 

During the study, air samples were taken in yellowcake packaging areas before, during, and after 

barrels of yellowcake were filled and sealed. Median air concentrations during the study ranged 

from 0.04 to 0.34 μg U/L (40 to 340 μg U/m3

). 
 

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Comparison of the above uranium mill results to Blockson operations cannot readily be made 

due to differences in operations and in quantities produced. Blockson produced a small quantity 

of yellowcake from phosphoric acid. Based on the reported masses produced, just a few drums, 

possibly up to 5 per month, would have been produced, assuming a nominal 1,000 pounds of 

yellowcake per drum. The mills processed uranium bearing ores that contained larger relative 

uranium concentrations, and in typically larger quantities. 

Although there are no air sampling results available for estimating intakes to Blockson workers, 

urinalysis results for uranium are available for Blockson workers, which is the preferred method 

of estimating internal dose received from exposure to uranium. Bioassay results are not 

available for individual claims submitted to NIOSH for dose reconstruction under EEOICPA, 

although the names on available results match a few workers whose job descriptions at Blockson 

are known. Default inhalation (or ingestion) intakes are presented below to be used for dose 

reconstructions when bioassay results are not available for a claim. 

The bioassay results were evaluated and assessed to provide intake rates that are favorable to 

claimants and are considered bounding for Blockson workers. This evaluation assumes two 

categories of workers. Workers that are considered to have the highest potential for exposure are 

categorized as production workers in this evaluation. This category includes operators who 

handled and packaged dried yellowcake routinely. Since there is no definitive data to 

differentiate exposure rates to production workers who are exposed to the highest concentrations 

and those exposed only intermittently, e.g., maintenance mechanics, all production workers are 

assumed to be exposed to the bounding concentration. The other category of exposure in this 

evaluation is administrative workers. The intake rates for administrative workers are based on 

the assumption that they would not be exposed to any significant degree in close proximity to the 

dried uncontained material, but could have been exposed to elevated levels of general area 

airborne uranium contamination on a continual basis. 

Records indicate that Blockson employed approximately 20 people in the uranium recovery 

operations. In 1951, prior to start up of operations, Blockson sent a document to the AEC that 

included a best estimate of manufacturing costs, with a breakdown of labor by category. 

Blockson projected the following personnel needs for the uranium operations: 2 Operators per 

shift, 1 Chemist per shift, 2 Daymen, 2 Mechanics (on average), 1 Clerk, 1 Development 

Chemist, and 1 Foreman. From the rate information given and total estimate it could be 

determined that Blockson assumed 4.2 shifts per day, which would allow for 24 hour operations 

every day of the week. This results in about 18 full time personnel and two part time personnel 

according to pre-operational estimates (Blockson 1951). 

In September 1953, after start up of operations, the AEC received a request for bioassay analysis 

services for “about twenty production workers engaged in uranium processing at Blockson 

Chemical Company” (AEC 1953). The first known sample results from these services were 

reported in April 1954 by the AEC New York Operations Office Health and Safety Laboratory 

(HASL). Subsequently, bioassay results on nine other occasions were reported by HASL 

through February 1958. Sample results are available for twenty five different workers. 
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The urinalysis records were found on reports from the AEC New York Operations Office, Health 

and Safety Division. One hundred twenty two sample results are available with the results 

ranging from 0 to 17 μg uranium per liter. The analyses were performed by fluorimetry. Of the 

twenty five workers, nineteen workers had multiple bioassay results, with six workers having a 

single sample result reported. The nineteen workers with multiple results were used to determine 

a distribution of inhalation intake rates. One of the nineteen workers had multiple samples over 

two different time periods with a two year break without urinalysis. For purposes of estimating 

worker intakes, that worker’s results were analyzed as two workers, resulting in a distribution of 

intake rates for twenty workers. The workers with only one sample result were not used, 

however, those six results ranged from 0 to 6 μg per liter, which fit in the distribution of bioassay 

results. 

Some of the names on the urinalysis reports have been matched with names of some workers 

whose jobs in Building 55 during uranium recovery work are known. These workers include two 

process operators, two chemists/analysts, and a supervisor. Some of these workers have been 

interviewed to confirm the work they performed in Building 55. Their bioassay results support 

the exposure assumptions in this evaluation, i.e., the highest exposed of those known workers 

was an operator whose job included drumming dried yellowcake, and the intake recommendations in this evaluation are favorable in comparison to the data for those workers. 

The International Commission on Radiological Protection (ICRP) recommends three material 

types for solubility of inhaled uranium, Types F, M, and S, based on the clearance rate from the 

lungs (ICRP 1994a). Various studies have shown that U3O8 closely corresponds to the clearance 

rate associated with material Type M. Some studies have also shown that high fired material can 

produce uranium compounds that clear more slowly from the lungs, i.e., indicative of material 

Type S (Rucker, et al. 2001). Type M uranium is the most appropriate lung solubility material 

type based on the process used for uranium extraction at Blockson. The U3O8 product was 

produced from wet phosphoric acid by filtering the precipitated uranium and then using a dryer 

to dewater the solids (Blockson 1953a). Calcination was done on phosphate rock prior to 

chemical processing and the intakes of highly insoluble Type S material from calciner could 

have occurred, as described in Section 3.2.1. After the calcined rock was digested in sulfuric 

acid and oxidized, no additional calcining or high firing of the uranium material was performed 

in the process used in uranium recovery in Building 55. Based on these processes and the results 

of various studies that have been summarized by Rucker, et al., Type M material is used to 

derive intakes from bioassay results. 

The Blockson bioassay results were reported in μg per liter, and were converted to μg per day by 

multiplying by a daily excretion rate of 1.4 liters, then converted to pCi/day by multiplying by 

0.677 pCi per μg of natural uranium. Individual worker intakes were determined using IMBAExpertTM by assuming a chronic inhalation intake of Type M uranium with parameters 

recommended by the ICRP (ICRP 1994b). 

For daily intake rate calculation purposes, intakes were assumed to have occurred beginning in 

the year sampled and ending with the last sample date. The results were given an absolute error 

of 1.0, which equally weights the sample results for fitting purposes in IMBA-Expert. Daily 

intake rates ranged from 6 to 76 pCi/day. The intake rate results fit well to a lognormal 
 

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distribution having a median value of 25 pCi/day with a geometric standard deviation of 2.1, as 

shown in Figure 3. The analysis of bioassay results with the assumption of 100% of the intake 

via inhalation is a claimant-favorable assessment of intakes from all pathways except when 

calculating dose to certain tissues of the gastrointestinal tract. Dose from ingestion is discussed 

below. 

Figure 3: Distribution of Uranium Intakes. 

5.0 

ln (intake, pCi/d) 4.0 

3.0 

2.0 

1.0 

0.0 

y = 0.7256x + 3.2181 

R2

 = 0.9639 

-2.50 -2.00 -1.50 -1.00 -0.50 0.00 0.50 1.00 1.50 2.00 2.50 

z score 

Production workers are assumed to have been continually exposed at the 95th percentile intake 

rate of 82 pCi/day, and administrative personnel are assumed to be exposed continually at the 

median intake rate of 25 pCi/day. The total uranium intakes are applied as 50% U-238 and 50% 

U-234, which allow for intakes of U-235. Intakes of long-lived radionuclides that are significant 

for internal dose are listed in Table 4a. These bounding intakes should be entered into IREP as 

constants. 

Table 4a: Inhalation Rate for Building 55.1,2,3 

Radionuclides Intake (pCi/d) 

Production workers 

Intake (pCi/d) 

Administrative workers 

U-238, Th-230, U-234, 

Pb-210, Po-210 

41 13 

Th-231, Pa-231, Ac-2274

 1.9 0.59 

Ra-226 1.9 0.59 

Th-232, Ra-228, Th-228 1.4 0.41 

1. Intake rates have been normalized to calendar days. 

2. Intakes are based on Type M lung solubility for materials likely to have been present in Building 55 

operations except for thorium, lead and polonium. Pb-210 is Type F, and Po-210 is Type F or M per 

ICRP 1994b. Thorium could have been Type M or Type S. Thorium and polonium solubility types 

must be selected based on the types that provide the largest dose to the organ or tissue of concern. 

3. See Table 3b for dose to tissues of the gastrointestinal tract. 

4. U-235 is allowed for in the U-238 and U-234 values. Values given are for radionuclides in the U235 chain. 

Workers also had the potential to ingest uranium from contact with contaminated surfaces or 

from eating or drinking in the area. When deriving intakes from the bioassay results, a chronic 

ingestion of uranium results in a higher dose to certain tissues of the gastrointestinal tract when 
 

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compared to the dose from the inhalation intakes described above. Therefore, intakes are 

presented in Table 4b based on the presumption that all the uranium in the workers urine was due 

to ingestion. Although inhalation is the most common mode of intake in a production facility the 

presumption of the ingestion pathway provides an upper bounding value for dose from ingestion. 

Since bioassay results are from intakes by all pathways a worker should be assigned Building 55 

intakes from inhalation or ingestion, not both. 

Table 4b: Ingestion rate for Building 55.1,2 

Radionuclide f1 Values4 

Intake (pCi/d) 

Production 

workers 

Intake (pCi/d) 

Administrative 

workers 

U-238, U-234, 

Th-230 

Pb-210 

Po-210 

0.02 

0.0005 

0.2 

0.1 

139 41 

Th-231, Pa-231, Ac-2273

 0.0005 6.4 1.9 

Ra-226 0.2 6.4 1.9 

Th-232, Th-228 

Ra-228 

0.0004 

0.2 

4.5 1.4 

1. Intake rates are normalized to units of calendar days. 

2. Ingestion intakes provide bounding dose to the stomach, small intestine, upper large intestine, lower large 

intestine, and colon. See Table 3a for estimating dose to all other tissues. 

3. U-235 is allowed for in the U-238 and U-234 values. Values given are for radionuclides in the U-235 chain. 

4. f1 values are from ICRP 1994b. 

3.2.3 Pilot Plant and Laboratory 

The location of the pilot plant operation that was built and used in 1951 and early 1952 has not 

been determined. As discussed in Section 2.2, the pilot plant operated to collect data on a 

production scale and operated at a much lower capacity than Building 55 and ran intermittently 

during developmental work. The plant tested methods determined from laboratory scale testing. 

The operations performed in the pilot plant were documented by Blockson (1953). Since this 

work involved the same source of radioactive materials as those in Building 55, except on a 

smaller scale and only operated intermittently for short periods, the doses modeled for Building 

55 workers is used to bound doses received by workers in the pilot plant operation and work in 

the laboratory. 

3.3 Radon Exposures 

Radon exposures to workers from uranium extraction work at phosphate plants have been 

evaluated for the NIOSH EEOCIPA dose reconstruction project (ORAUT 2006). For 

reconstructing lung doses, all workers at Blockson during the AEC operational period are to be 

assigned an exposure at the 95th percentile of 0.112 WLM (working level month) per year due to 

radon progeny. Dose to other organs are assigned as alpha doses. Table 5 lists doses from radon 

exposures. 
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Table 5: Radon Exposures. 

Dose component Annual dose/exposure1

 Distribution 

Radon progeny 0.112 WLM (lungs only) Constant value 

Radon progeny ET1 and ET2 tissues2 Constant value 

Radon gas 0.002 rem alpha (nonrespiratory tract tissues only) Constant value 

1. Exposure and dose values from ORAU 2006. Values are normalized for a 365 day year. 

2. ET1 and ET2 doses are to be applied as alpha dose and calculated from WLM values using conversion 

factors in NIOSH 2006b. 

4.0 External Dose 

Because phosphate rock contains Naturally Occurring Radioactive Material (NORM), work 

involving phosphates potentially exposes workers to radioactivity. External dosimetry data is 

not known to exist for Blockson workers, and data capture efforts for the EEOICPA dose 

reconstruction project have not found any direct radiation survey results for the Blockson facility 

during the uranium extraction operational period. Exposure from operations outside of Building 

55 has been considered based on studies of doses received by workers at similar plants, which 

are discussed in Section 4.1. Those studies did not include dose received from uranium 

extraction operations. Therefore, source term information has been used to estimate external 

doses to workers involved in uranium extraction operations as described in Section 4.2. 

4.1 External Dose from Existing Operations 

TLD data, evaluated by the Florida Institute of Phosphate research, (FIPR 1998 tables 14-18) 

indicate that workers involved in various phases of phosphate rock handling, crushing and 

phosphoric acid liquor production received between 10 and 210 mrem (maximum) per year (at 

other wet process phosphoric acid operation sites). Eighty percent of the workers monitored 

received <10 mrem per year, 15% received >10 but <99 mrem/year and less than 2% received 

greater than 100 mrem/yr with 210 mrem (or 0.210 rem) being the maximum recorded dose. 

Radiological survey data available in FIPR 1998, Idaho 1978, and 1996 post site operations 

surveys of the Blockson plant in Joliet all support the TLD results (OLIN 1996). ORAU 2006 

estimates an upper bound external dose to workers at wet process phosphate plants at 0.220 

rem/yr. The doses would include external exposures to the radium bearing phosphogypsum that 

is filtered out during phosphoric acid production, but would not have included doses that have 

been received due to concentration of uranium, such as Blockson did in Building 55. The 0.220 

rem/yr bounding external dose from other phosphate facilities are lower than the modeled 

external doses from Building 55 discussed in Section 4.2. Therefore, Building 55 modeled doses 

are considered to be an upper bound of the external dose from the entire Blockson facility. 

4.2 External Dose from Uranium Extraction Operations 

For the purpose of dose reconstruction from uranium extraction work in Building 55, it is 

assumed that there was a potential for external dose due to exposure from submersion in air, 

from exposure to contaminated surfaces and from exposure to barrels of the final concentrated 

uranium product. 
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Based on radioactivity air concentrations derived from the daily intake rates discussed in Section 

3.2.2, and the dose coefficients from Federal Guidance Report No. 12 (EPA 1993), the external 

dose from submersion in air contaminated with radioactive dust is insignificant in comparison to 

the favorable evaluation of other dose components and, therefore, is not included in dose 

reconstructions. 

4.2.1 Source Term 

Although uranium and the short lived progeny of U-238 are the most abundant radionuclides in 

Building 55, the presence of other radionuclides that may have followed the uranium to Building 

55 can produce significant external doses at much lower concentrations. Section 3.1 discussed 

the various radionuclides that are of significance at Blockson. For modeling external dose in 

Building 55, assumptions were made to allow for radionuclides that may have been present as 

contaminants in Building 55. Allowances were also made to consider ingrowth of some 

radionuclides due to radioactive decay of parent radionuclides. Those assumptions are included 

in the the Table 1 source term assumptions have been used to model bounding external doses in 

Building 55. 

Clegg and Foley (1958) state that freshly separated yellowcake has a very low gamma emission 

rate; therefore external radiation is of no particular concern at this stage of the process. 

However, due to ingrowth of daughter radionuclides in the yellowcake, the radiation levels 

increase for several months following production (NRC 2002b). 

For accumulations of processed yellowcake dust, the surface beta dose rate from U-238 

daughters is negligible just after separation, but rises steadily until Pa-234m and Th-234 reach 

equilibrium concentrations. After a few months, the beta surface dose rate is about 150 mrem/hr 

(NRC 2002a). Figure 4 shows the rise in beta dose rate during 100 days after separation from 

ore. 

Figure 5 shows that the beta dose rate from the surface of yellowcake decreases rapidly as a 

function of distance from the surface. Rapid decrease in the beta dose rate with distance, and the 

shielding afforded by shoes and clothing, reduces dose from beta radiation, particularly from 

yellowcake deposited on floors. 
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Figure 4. Beta Dose Rate on the Surface of Yellowcake. [Reproduced from 

NRC 2002a] 

Figure 5. Beta Dose Rate from Yellowcake Separated from Ore for More 

Than 100 Days as a Function of Distance from the Surface. 

[Reproduced from US NRC 2002a] 
 

 

 

 

 

 

 

 

 

 

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4.2.2 Exposure from Drums of Uranium 

MCNPX (version 2.5.0) was used to determine the dose rate per Curie of 238U regardless of the 

actual activity in the drum. This was later adjusted for actual source activity to compare actual 

dose rates. All radionuclide concentrations were calculated based on the ratio to 238U for 

determination of the number of photons and electrons per decay of 238U. For the purposes of this 

evaluation, branching ratio adjusted equilibrium was assumed. A significant amount of progeny 

was included, including portions of the 232Th chain, based on the discussion in Sections 3.1 and 

4.1.2. The details are provided in Table 6. 

Table 6: Radioactivity Ratios for External Dose Modeling. 

Activity 

Concentration 

in aged 

Uranium Metal 

(Bq/g) 

Relative 

activity 

concentration 

(normalized to 

238U)

238U 12200 1 

234Th 1 

234mPa 1 

234Pa 0.0016 

234U 1 230Th 1 226Ra 0.047 222Rn 0.047 218Po 0.047 214Pb 0.047 214Bi 0.047 214Po 0.047 210Pb 1 210Bi 1 210Po 235U 0.045492 231Th 0.045492 231Pa 0.045492 227Ac 0.045492 227Th 0.044855 223Fr 0.000628 223Ra 0.045492 

219Rn 0.045492 215Pa 0.045492 211Pb 0.045492 211Bi 0.045492 211Po 0.000126 207Tl 0.045366 
 

 

 

 

 

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Table 6, cont. 232Th and progeny 232Th 0.033 228Ra 0.033 228Ac 0.033 228Th 0.033 224Ra 0.033 220Rn 0.033 216Po 0.033 212Pb 0.033 212Bi 0.033 208Tl 0.01188 212Po 0.02112 

Dose 30 cm from drum 

The dose rate was determined at 77.9 cm above the ground, 30 cm from the edge of the drum for 

both the photon and beta emissions of natural uranium and its progeny. ICRP Publication 74 

Table A.1 was used to convert the photon flux to units of air kerma. 

The effect of density of the drummed uranium concentrate on the modeled dose rate was 

evaluated. The effective density of the drummed material was assumed to be variable up to 

6.7 g cm-3 . Based on these results, the effective density of drummed U3O8 concentrate makes 

little difference in the calculated dose rates from 1 to 6.7 g cm-3 . Results are provided in Table 7 

and Figure 6. 

Table 7: Dose Rates from Drums of Yellowcake. 

Density of 

U3O8 (g cm-3) 

Activity of U in 

drum (Ci) 

Photon 

emission 

dose(rad/hr) 

Bremsstrahlung 

dose (rad/hr) 

Total dose rate 

at 30 cm 

(rad/hr) 

1 6.242E-02 5.12E-03 4.08E-04 5.53E-03 

2 1.248E-01 5.62E-03 4.39E-04 6.06E-03 

4 2.497E-01 5.97E-03 4.54E-04 6.42E-03 

6 3.745E-01 5.99E-03 4.38E-04 6.43E-03 

6.7 4.182E-01 6.02E-03 4.52E-04 6.47E-03 

*The drum begins to noticeably impact the dose rates at low material concentration. 
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Figure 6: Effect of Density on Dose Rate at 30 cm from Drum of Yellowcake. 

density (g cm-3) 

0 1 2 3 4 5 6 7 

rad hr-1 

0.000 

0.001 

0.002 

0.003 

0.004 

0.005 

0.006 

0.007 

photons 

bremsstrahlung 

total 

Cumulative energy distribution results are shown in Figure 7.Assignment of 10% dose from 30­

250 keV photons and 90% dose from >250 keV photons was verified to be favorable to 

claimants as only approximately 1% of the total dose comes from photons less than 300 keV 

(originating from both bremsstrahlung and photon emissions) in a drum with density of 4 g/cm3 

Figure 7: Cumulative Dose Rate from Photons and Bremsstrahlung at 30 cm 

from a Drum Containing U3O8 (density 4 g/cm3

). 

cumulative rad/hr 

0.000 

0.001 

0.002 

0.003 

0.004 

0.005 

0.006 

0.007 

. 

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 

Photon Energy (MeV) 
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The air kerma values were converted to annual organ doses by assuming a worker’s exposure 

time was lognormally distributed. The median exposure time was determined by assuming all 

workers were working eight hours per day, one day per week at a distance of 1 foot (30 cm) from 

the drum. This was normalized to 400 hours per work year. The 95th percentile exposure time 

was determined by assuming the worker spent 2000 hours per year at a distance of 1 foot from 

the drum. This results in a whole body dose distribution with a median value of 2.572 rad per 

year with a geometric standard deviation of 2.7. 

To calculate organ doses for use in the NIOSH Interactive RadioEpidemiological Program 

(NIOSH-IREP) Monte Carlo methods were used to multiply the whole body dose and energy 

split times the triangular organ dose conversion factors for kerma to organ dose found in NIOSH 

External Dose Reconstruction Implementation Guideline (NIOSH 2006a). The results are annual 

doses that are lognormally distributed. The results are in Table 8. For skin, air kerma values 

were multiplied by an organ dose conversion factor of 1.0. 

Table 8: Annual Organ Dose from Exposure to Drums of Yellowcake. 

30-250 keV photons >250 keV photons 

Organ dose, rem GSD dose, rem GSD 

Bladder 0.307 2.7 2.567 2.7 

RBM 0.132 2.8 1.973 2.7 

Bone Surface 0.349 2.7 2.161 2.7 

Breast (female) 0.352 2.7 2.707 2.6 

Colon 0.243 2.8 2.432 2.6 

Esophagus 0.145 2.9 2.059 2.6 

Eye 0.361 2.7 2.557 2.7 

Ovaries 0.211 2.8 2.303 2.7 

Testes 0.406 2.7 2.771 2.7 

Liver 0.249 2.8 2.442 2.7 

Lung 0.230 2.8 2.361 2.7 

Remainder 0.199 2.9 2.193 2.7 

Stomach 0.309 2.7 2.608 2.6 

Thymus 0.357 2.8 2.644 2.7 

Thyroid 0.389 2.7 2.817 2.7 

Uterus 0.230 2.8 2.246 2.7 

Skin 0.257 2.7 2.315 2.7 

4.2.3 Exposure from Contaminated Surfaces 

Estimates of external dose from surfaces contaminated with uranium have been performed. The 

95th percentile intake rates from inhalation were used to derive a U-238 airborne concentration 

of 4.3 pCi/m3 . A terminal settling velocity of 0.00075 m/s was used as an estimate of the 

velocity of deposition to surfaces in the building. The value is within the range of deposition 

velocities measured in various studies (NRC 2002b). It was assumed that uranium settled on 
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plant surfaces at a steady state 24 hours per day for 365 consecutive days with no cleanup or 

removal of contamination. 

The estimated surface contamination results for U-238 and associated radionuclides were 

multiplied by the Dose Coefficients for Exposure to Contaminated Ground Surface found in 

Table III.3 of Federal Guidance Report No. 12 (EPA 1993). With the exception of dose to the 

skin, the annual dose for all organs is insignificant in comparison to the favorable assignment of 

dose from drums of yellowcake. Beta dose to the skin is discussed in Section 4.4. 

Although the modeled external doses are assumed to be insignificant based on airborne uranium 

concentrations there could have been localized spots with greater potential for exposure. Direct 

reading radiation results are available from surveys done from March to November 1978 by 

Argonne National Laboratory (ANL) (DOE 1983). 

The ANL survey was performed throughout Building 55, including plant surfaces, tanks, pipes, 

and other process equipment. The report estimated that some areas were not accessible, but that 

they had surveyed an estimated 95% of the floors and 90% of the walls. The surveys included 

contamination surveys. A dose rate was taken at contact and at 1 meter on all 63 spots in which 

contamination was detected. The dose rates at 1 meter on 7 of the 63 “hot” spots ranged from 

0.04 mR/hr to 0.2 mR/hr. The other 56 spots had 1 meter dose rates indistinguishable from 

background. The reported background dose rate on the instrument used was between 0.02 mR/hr 

and 0.03 mR/hr. The results of the 7 spots with measurable 1 meter dose rates included the 

background dose rates. From a review of the survey map and results it seems improbable that a 

worker could be significantly exposed above the background rate of 0.03 mR/hr for significant 

time. However, in the absence of individual dosimeter data, whole body dose rates are modeled 

by a lognormal distribution by assuming a worker was exposed to the 0.03 mR/hr rate for 2,000 

hours per year, which results in an annual exposure of 60 mR, or 0.06 R. To allow for 

uncertainty this value is applied as the median of a lognormal distribution. The geometric 

standard deviation is 3.2, which was determined by assuming that the 95th percentile dose rate is 

equal to the maximum observed result of 0.2 mR/hr. 

The 1978 survey suggests that either the facility was only contaminated in localized spots, or it 

indicates that some areas of surface contamination were not subject to the same contamination 

depletion rate that would be expected from natural processes and from general cleaning and 

weathering that would have occurred since 1962. The 7 spots were on small areas of floor 

(reported to be about 0.5 m2

 each), on a pipe inlet, and on a spot on a machine. Additionally, 

since the normal non-AEC related operations at Blockson produced byproduct that contained 

small amounts of uranium and daughter products, it is unknown if that would have contributed to 

the contamination of the facility after AEC operations ended in 1962. Based on the above 

considerations and the absence of more data, an assumption that is favorable to claimants is 

made for deriving dose rates prior to 1978. It is assumed that the few spots in the facility with 

measurable dose rates were representative of the entire facility from the operational period to the 

present or until the end of the workers’ EEOICPA covered employment. The deep dose rates 

during the operational period are adequately bounded by the assumptions made in deriving doses 

from drums of yellowcake. Therefore, dose from contaminated surfaces is only assigned during 

the residual contamination period. 
 

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4.2.4 Beta Dose 

It is assumed that there was a potential to receive a shallow dose from exposure to open drums of 

yellowcake during drum loading and sealing. According to Figure 5 the dose rate at 1 foot from 

the surface of aged yellowcake is between 1 and 2 mrem/hour. It is assumed that the production 

workers spent 8 hours per week, 50 weeks per year, at 1 foot from the surface of aged 

yellowcake at a dose rate of 2 mrem/hour. To allow for uncertainty, the time of exposure was 

assumed to be lognormally distributed with the 95th percentile exposure time assumed to 40 

hours per week, 50 weeks per year. This results in an upper shallow dose of 0.8 rem per year 

with a geometric standard deviation of 2.7. The 0.8 rem per year was adjusted to allow for beta 

dose from other radionuclides that are assumed to be present in the uranium per the ratios in 

Table 6. The relative activity of each radionuclide was applied to Federal Guidance Report No. 

12 dose conversion factors for skin for exposure to contaminated surfaces. Those factors 

indicate that U-238, Th-234, Pa-234m, Pa-234, and U-234 account for about 66% of the skin 

dose for the Table 6 ratios. Adjusted beta dose is provided in Table 9. The calculated beta doses 

have not been reduced to allow for doses to areas of the skin that are typically covered by 

clothing that reduces beta dose to the skin. 

It is also assumed that there was a potential for workers to receive a shallow beta dose to the skin 

contaminated with yellowcake. Skin contamination from contaminated air was estimated by 

using the measured skin deposition velocity of 4-μm particles to skin of 0.012 m/s (Andersson et 

al. 2002; Fogh 1999), assuming that the material was deposited on the skin for an entire 8-hour 

shift. The dose is negligible when compared to the shallow dose estimate from exposure to a 

drum of aged yellowcake discussed above and the estimated dose from contact with 

contaminated surfaces described below. 

Skin dose has also been considered from contact with contaminated work clothing. Average 

dose data from contaminated clothing at Mallinckrodt indicate levels of 1.5 mrem/hour (AEC 

1958b). The Mallinckrodt dose rate is used as a bounding condition for Blockson because 

Mallinckrodt handled materials of similar radiological constituents, but in larger quantities and 

with a higher radioactive material content. It is assumed that the workers were exposed to that 

level for 1000 hours per year, which is considered an upper bound condition. This results in a 

dose to the skin of 1.5 rem per year. Doses are applied as electrons > 15keV. 

Former workers have said that the filtering operation in Building 55 exposed their hands directly 

to the filter cake that contained the uranium. They said they wore gloves for this work, but 

sometimes would have to take the gloves off and use their bare hands to remove the product 

from the filters (OCAS 2007a). Dose for this activity has been estimated for the hands and 

forearms. Yellowcake concentrations in the product delivered to the AEC was estimated to be 

40% - 60%. To determine an upper bound dose, an estimate was made of shallow dose to the 

hands based on direct contact with pure yellowcake. Surface dose rates on yellowcake have been 

reported to be about 203 mrad per hour (DOE 2000). The time of direct contact has been 

assumed to be 2 hours per week, 50 weeks per year during the operational period. The 66% 

adjustment factor discussed above was applied to allow for non-uranium contaminants. This 
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results in an annual dose of 30 rem to the hands and forearms. The dose applies to filter 

operators only and are applied as electrons > 15keV. 

Table 8 contains a summary of shallow dose from electrons. 

Table 9: Beta Dose to Skin. 

Dose component 

Beta dose, E>15 keV 

Annual dose1

 Distribution 

Dose from drums of 

yellowcake 

1.2 rem per year Lognormal, GSD=2.7 

Dose from contaminated 

clothing 

1.5 rem per year Constant 

Dose to hands and 

forearm from contact 

with yellowcake 

30 rem per year 

(filter operators only) 

Constant 

1. Beta dose is applicable for the operational period only. 

4.3 Occupational Medical Dose 

Dose from occupationally required medical X-rays has also been considered and assumed to 

have occurred, although no information has been found to indicate that Blockson or the AEC 

required X-rays of the workers. For the AEC operational period at Blockson, employees are 

assumed to have received an annual chest X-ray. Organ doses are listed in Table 10 and are 

based on an assumed Posterior-Anterior (PA) exposure with minimal collimation. Dose values 

are reproduced from Table 6-5 of “Dose Reconstruction from Occupationally Related Diagnostic 

X-ray Procedures” (ORAU 2005.). The annual doses are applied as dose from 30-250 keV 

photons using the values in Table 10 as the mean of a normal distribution with a 30% standard 

deviation. 

Table 10: Annual Organ Dose from Medical X-rays. 

Organ 

Annual dose, rem 

photon 30-250 keV 

Thyroid 3.48E-02 

Eye/brain 6.40E-03 

Ovaries 2.5E-02 

Liver/gall bladder/spleen 9.02E-02 

Urinary bladder 2.5E-02 

Colon/rectum 2.5E-02 

Testes 5.0E-03 

Lungs (male) 8.38E-02 

Lungs (female) 9.02E-02 

Thymus 9.02E-02 

Esophagus 9.02E-02 

Stomach 9.02E-02 

Bone surfaces 9.02E-02 
 

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Table 10, cont. 

Remainder 9.02E-02 

Breast 9.80E-03 

Uterus 2.5E-02 

Bone marrow (male) 1.84E-02 

Bone marrow (female) 1.72E-02 

Skin 2.70E-011 

1. Skin dose is for skin in the primary beam where the beam enters the body. 

5.0 Dose from Residual Contamination 

The whole body median dose rate of 0.060 R/year derived from the discussion in Section 4.2.3 

above was used to calculate various organ doses. Photon dose is split 10% 30-250 keV and 90% 

>250 keV ranges. The Roentgen to organ dose conversion factors for isotropic geometry in the 

External Dose Reconstruction Implementation Guideline (NIOSH 2006a) were used to estimate 

organ doses. Skin doses were calculated using an organ dose conversion factor of 1.00. Results 

are in Table 11. 

Table 11: Annual Dose from Residual Contamination.1 

Photons 

30-250 keV 

Photons 

>250 keV 

Organ dose, rem dose, rem 

Bladder 3.2E-03 3.5E-02 

RBM 3.3E-03 3.6E-02 

Bone Surface 5.6E-03 3.7E-02 

Breast (female) 4.2E-03 4.0E-02 

Colon 3.1E-03 3.4E-02 

Esophagus 3.0E-03 3.5E-02 

Eye 4.5E-03 4.1E-02 

Ovaries 3.0E-03 3.4E-02 

Testes 3.8E-03 3.7E-02 

Liver 3.4E-03 3.6E-02 

Lung 3.8E-03 3.8E-02 

Remainder 3.3E-03 3.6E-02 

Stomach 3.4E-03 3.6E-02 

Thymus 3.7E-03 3.7E-02 

Thyroid 3.8E-03 3.9E-02 

Uterus 2.9E-03 3.3E-02 

Skin1 6.0E-03 5.4E-02 

1. For dose reconstructions, annual doses are applied as lognormal 

distributions with a GSD of 3.2. 

Uranium inhalation intakes during the residual contamination period have been derived from the 

operational period intakes and from estimated airborne radioactivity derived from the 1978 

FUSRAP survey. While the uranium recovery operations could result in high localized air 

concentrations, air concentrations from resuspension of residual contamination would be more 
 

 

 

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consistent throughout the area. Interviews with former workers indicate that housekeeping was 

performed regularly to reduce build up of material on the floors (OCAS 2007a). After cessation 

of uranium recovery work the main source of contamination (precipitated and dried yellowcake) 

was no longer present. Therefore, the derived median U-238 inhalation rate of 13 pCi/day is 

used as the inhalation intake rate of U-238 at the start of the residual contamination period on 

April 1, 1962. Thereafter, airborne radioactivity from resuspension of contamination in the 

facility and corresponding intakes are assumed to decrease according to an exponential model 

described below. 

Extensive radiological surveys were performed starting in March 1978. Airborne radioactivity 

concentration at 16 years post operation was estimated by assuming that the facility was 

uniformly contaminated at the level of the maximum alpha smear result of 640 dpm/100 cm2

 as 

reported in the 1978 survey (DOE 1983). This value was multiplied by a resuspension factor of 

1E-06 m-1 (NRC 2002c), which results in an estimated maximum residual alpha air concentration 

of 0.029 pCi/m3 . Application of a breathing rate of 9.6 m3/day results in a potential alpha 

particle inhalation of about 0.28 pCi/day at the time of the March 1978 survey. The assumption 

of are equal amounts of U-238, U-234, and Th-230 and that they account for all alpha particles, 

results in a U-238 intake of about 0.092 pCi/day. The 1962 and 1978 daily intake rates were 

used to estimate the annual intakes from the following equation. 

It = I 0 * e−λ t 

where: 

It = daily intake rate at time t 

t = time (days) since April 1, 1962 

I0 = daily intake on April 1, 1962 

e = base of the natural logarithms 

λ = exponential constant 

The derived U-238 median intake of 13 pCi/day on April 1, 1962, was substituted for I0. The 

0.092 pCi/day derived U-238 intake from the 1978 survey was substituted for It on April 1, 1962. 

The time between April 1, 1962, and April 1, 1978 is 5844 days. This resulted in the following 

equation to calculate the exponential constant λ. 

0.092 pCi 13 pCi −λ*5844 d = * e

d d 

The constant λ was determined to be 0.000847 day-1 . 

Average daily inhalation intake rates for each year from 1962 through 1996 were then calculated 

and are given in Table 12a below. The methods used for derivations of these intakes are 

considered bounding, and the corresponding annual doses are considered constants for purposes 

of dose reconstruction. 
 

 

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Table 12a: Inhalation Intake Rate from Residual Contamination.1,2,3,4 

U-238, Th-230, 

U-234, Pb-210, 

Po-210 

Th-231,5 Pa-231, 5 

Ac-227, 5 Ra-226 

Th-232, Ra-228, 

Th-228 

Year pCi/day pCi/day pCi/day 

1962 12 0.54 0.38 

1963 8.9 0.42 0.29 

1964 6.5 0.31 0.22 

1965 4.8 0.22 0.16 

1966 3.5 0.16 0.12 

1967 2.6 0.12 0.085 

1968 1.9 0.089 0.062 

1969 1.4 0.065 0.046 

1970 1.0 0.048 0.034 

1971 0.75 0.035 0.025 

1972 0.55 0.026 0.018 

1973 0.40 0.019 0.013 

1974 0.29 0.014 0.010 

1975 0.22 0.010 0.007 

1976 0.16 0.0075 0.005 

1977 0.12 0.0055 0.0038 

1978 0.085 0.0040 0.0028 

1979 0.063 0.0029 0.0021 

1980 0.046 0.0022 0.0015 

1981 0.034 0.0016 0.0011 

1982 0.025 0.0012 0.0008 

1983 0.018 0.0009 0.0006 

1984 0.013 0.0006 0.0004 

1985 0.0098 0.0005 0.0003 

1986 0.0072 0.0003 0.0002 

1987 0.0053 0.0002 0.0002 

1988 0.0039 0.0002 0.0001 

1989 0.0028 0.0001 0.0001 

1990 0.0021 0.0001 0.0001 

1991 0.0015 0.0001 0.0001 

1992 0.0011 0.0001 0.00004 

1993 0.0008 0.00004 0.00003 

1994 0.0006 0.00003 0.00002 

1995 0.0004 0.00002 0.00001 

1996 0.0003 0.00002 0.00001 

1. Inhalation intakes are not assigned for calculating dose to the stomach, small 

intestine, upper large intestine, lower large intestine, or colon. See Table 12b for 

ingestion intakes for those tissues. 

2. 1962 intakes from residual contamination start on April 1. Building 55 was 

demolished in 1996. 
 

 

 

 

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3. Intakes are based on Type M lung solubility for materials likely to have been present 

in Building 55 operations except for thorium, lead and polonium. Pb-210 is Type F, 

and Po-210 is Type F or M per ICRP 1994b. Thorium could have been Type M or 

Type S. Thorium and polonium solubility types must be selected based on the types 

that provide the largest dose to the organ or tissue of concern. 

4. See Table 11b for dose to tissues of the gastrointestinal tract. 

5. U-235 is allowed for in the U-238 and U-234 values. Values given are for 

radionuclides in the U-235 chain. 

Inhalation of uranium is considered to be proportional to the airborne concentration. For the 

residual period this is a function of the amount of loose contamination present on plant surfaces. 

Likewise, the potential for ingestion of uranium is a function of the amount of loose 

contamination present on plant surfaces. This indicates a relationship between airborne 

concentration and ingestion of material. Bounding ingestion intakes were evaluated and 

discussed above for the AEC operational period. Based on the median intake rate, that 

evaluation indicates that doses to the gastrointestinal tract are bounded by an ingestion shown in 

Section 3.2.2 (administrative scenario) , which is based on the evaluation of bioassay results 

under the assumption that no inhalation occurred. 

To estimate bounding intakes for ingestion during the residual contamination period the daily 

ingestion rate is reduced at the same rate as the inhalation intakes described above. This results 

in the following expression to determine ingestion intakes during the residual contamination 

period. 

t I = d 

pCiI 0 td

e 

0.000847 *

1

* − −

where: 

It = daily ingestion intake rate at time t 

I0 = bounding ingestion intake on April 1, 1962 

t = days since April 1, 1962 

The above equation was used to derive average daily ingestion intake rates for each year between 

1962 and 1978. Results are presented in Table 11b. The ingestion intake rate is based on 

bounding dose to the stomach, small intestine, upper large intestine, and lower large intestine. 

The modeled ingestion intake rate assumes that all material assimilated from contaminated 

surfaces is via the ingestion pathway. Therefore, inhalation intakes are not assigned for these 

tissues. The doses are considered constants for dose reconstruction purposes. 
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Table 12b: Ingestion Intake Rate from Residual Contamination1

 U-238, Th-230, 

U-234, Pb-210, 

Po-210 

Th-231,5 Pa-231, 5 

Ac-227, 5 Ra-226 

Th-232, Ra228, Th-228 

Year pCi/day pCi/day pCi/day 

1962 37 1.8 1.2 

1963 29 1.3 0.94 

1964 21 1.0 0.69 

1965 15 0.72 0.51 

1966 11 0.53 0.37 

1967 8.3 0.39 0.27 

1968 6.1 0.29 0.20 

1969 4.5 0.21 0.15 

1970 3.3 0.15 0.11 

1971 2.4 0.11 0.08 

1972 1.8 0.08 0.06 

1973 1.3 0.06 0.043 

1974 1.0 0.045 0.031 

1975 0.70 0.033 0.023 

1976 0.51 0.024 0.017 

1977 0.38 0.018 0.012 

1978 0.28 0.013 0.0091 

1979 0.20 0.010 0.0067 

1980 0.15 0.0070 0.0049 

1981 0.11 0.0051 0.0036 

1982 0.080 0.0038 0.0026 

1983 0.059 0.0028 0.0019 

1984 0.043 0.0020 0.0014 

1985 0.032 0.0015 0.0010 

1986 0.023 0.0011 0.0008 

1987 0.017 0.0008 0.0006 

1988 0.013 0.0006 0.0004 

1989 0.0092 0.0004 0.0003 

1990 0.0067 0.0003 0.0002 

1991 0.0050 0.0002 0.0002 

1992 0.0036 0.0002 0.0001 

1993 0.0027 0.0001 0.0001 

1994 0.0020 0.0001 0.0001 

1995 0.0014 0.0001 0.0000 

1996 0.0011 0.0000 0.0000 

1. Ingestion intakes are assigned only when calculating dose to the stomach, small 

intestine, upper large intestine, lower large intestine, and colon. No inhalation intake 

is assigned for these tissues. 

2. 1962 intakes from residual contamination start on April 1. 

3. The f1 values are the same as those in Table 4b. 
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Radon exposure from residual contamination is assumed to have occurred; these values drop 

over time to be lower than that received on site from natural radioactivity or other operations on 

site. A gradual decline in radon in Building 55 from AEC activities is also assumed to have 

occurred coinciding with the end of operations in 1962. The decline is assumed to be at the same 

rate as the particulate contamination discussed above, starting with the bounding annualized 

operational exposure of 0.112 WLM specified in Table 5. The annual radon exposures are listed 

in Table 12. 

Table 13: Annual Radon Exposure from Residual Contamination. 

Lungs Non-respiratory 

tract tissues 

Year WLM/yr1

 rem/yr, alpha 

1962 9.95E-02 0.002 

1963 7.63E-02 0.002 

1964 5.62E-02 0.001 

1965 4.11E-02 0.001 

1966 3.02E-02 0.001 

1967 2.21E-02 0.001 

1968 1.63E-02 0.0004 

1969 1.19E-02 0.0003 

1970 8.75E-03 0.0002 

1971 6.42E-03 0.0002 

1972 4.73E-03 0.0001 

1973 3.46E-03 0.0001 

1974 2.54E-03 0.0001 

1975 1.86E-03 <0.0001 

1976 1.37E-03 <0.0001 

1977 1.00E-03 <0.0001 

1978 7.37E-04 <0.0001 

1979 5.41E-04 <0.0001 

1980 3.98E-04 <0.0001 

1981 2.91E-04 <0.0001 

1982 2.14E-04 <0.0001 

1983 1.57E-04 <0.0001 

1984 1.15E-04 <0.0001 

1985 8.45E-05 <0.0001 

1986 6.20E-05 <0.0001 

1987 4.55E-05 <0.0001 

1988 3.35E-05 <0.0001 

1989 2.45E-05 <0.0001 

1990 1.80E-05 <0.0001 

1991 1.32E-05 <0.0001 

1992 9.71E-06 <0.0001 
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Table 13, cont. 

1993 7.11E-06 <0.0001 

1994 5.22E-06 <0.0001 

1995 3.83E-06 <0.0001 

1996 2.82E-06 <0.0001 

1. ET1 and ET2 doses are to be calculated from WLM values using 

conversion factors in NIOSH 2006b. 

6.0 Dose Reconstruction Summary 

As indicated in the discussion in the previous sections, Blockson processed materials that 

contained Naturally Occurring Radioactive Materials (NORM), with the uranium work for the 

AEC being a byproduct of the existing operations (Blockson 1958). Therefore, employees had 

the potential to receive radiation dose to some extent both before and after the uranium 

extraction operations in Building 55. During the 1951 through 1962 AEC operational period 

industrial doses workers may have received in the covered facilities are included in dose 

reconstructions. Therefore, bounding internal and external doses have been considered for the 

entire Blockson site during those years. 

This document evaluated two possible scenarios to determine bounding doses received by 

workers during the AEC operational period: dose inside Building 55 to bound doses received by 

workers involved with uranium extraction work, and dose outside of Building 55 to bound doses 

received elsewhere at the plant. The claimant-favorable modeling of external dose in Building 

55 bounds dose from the entire site. 

For internal dose the bounding intakes from calcination of phosphate rock listed in Table 3 are 

lower than the intakes from Building 55 listed in Tables 4a and 4b. However, the intakes from 

the calciner may result in higher dose to some organs, e.g., lungs, due to solubility types likely to 

be present. Therefore, bounding internal doses for the site should be selected based on the values 

that result in the higher organ doses, whether it is from Table 3 or Table 4a and 4b. 

For EEOICPA dose reconstruction purposes, exposure starts in March 1951, or the first date the 

employee has covered employment at Blockson, whichever is later. The end of the operational 

period is March 31, 1962. Residual contamination doses start on April 1, 1962. Building 55 was 

demolished in 1996 and residual doses from that building should be assigned through that year. 

Table 14 contains a summary of the tables in this document to use for dose reconstructions. 
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Table 14: Dose Reconstruction Summary Table 

Internal Dose Dates Dose Tables 

Operational Period March 1951 through 

March 1962 

Table 3 or Table 4a/4b 

Table 5 

Residual Period April 1962 through 

December 1996 

Table 12a/12b 

Table 13 

External Dose Dates Dose Tables 

Operational Period March 1951 through 

March 1962 

Table 8 

Table 9 

Residual Period April 1962 through 

December 1996 

Table 11 

Medical X-rays Annual 1951 through 1962 Table 10 
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7.0 Attributions and Annotations 

Thomas Tomes of NIOSH/OCAS was the principle author of this document. He also performed 

the internal dose evaluations and evaluated dose from residual contamination. 

Sam Glover, PhD, of NIOSH/OCAS performed the MCNP external dose modeling based on 

assumptions stated in the document. 

Some site information and external dose evaluation was retained from the original version of the 

site profile developed by the ORAU team under contract to NIOSH. Jeri Anderson, PhD, was 

the subject expert of the original version developed by ORAU. This revision incorporated 

additional information from former Blockson workers that was not previously available. Newly 

obtained data from DOE archives and from the Olin Corporation was also included in this 

revision. 

Chemical modeling was added to account for potential doses from the various progeny present 

and to assess its potential impact on worker dose. Supporting information and evaluation was 

provided by George Vargo, PhD, of the ORAU team, who contracted Alan Elzerman, PhD, of 

Clemson University to evaluate certain Blockson chemical processes. The conclusions and 

recommendations in this document are those of the principal author. The conclusions have been 

reviewed, and comments by peers and OCAS management have been incorporated. 
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8.0 References 

AEC (U. S. Atomic Energy Commission) 1951, Contract for Uranium Production, Contract No. 

AT- (49-1)-611, October 18, 1951 [SRBD Ref ID 16328]. 

AEC 1953, letter from James A. Barr, Jr., Assistant Director, Domestic Production, Division of 

Raw Materials, to Dr. Richard Fahrner, Medical Consultant, Blockson Chemical Company, 

September 14, 1953 [SRBD Ref ID 9728, p. 16]. 

AEC 1955a, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates, Office of Domestic Procurement,” Washington D.C., August 31, 1955 [SRDB Ref 

ID 22264]. 

AEC 1955b, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

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Ref ID 22264]. 

AEC 1955c, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

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ID 22264]. 

AEC 1955d, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

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AEC 1955e, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates,” Office of Domestic Procurement, Washington D.C., December 1955 [SRDB Ref 

ID 22264]. 

AEC 1958a, letter from R. H. Kennedy, Division of Raw Materials, to C. L. Karl, Area Manager, 

Fernald Area Office, “Official Sampling of Blockson Concentrate,” February 27, 1958 [SRDB 

Ref ID 29493]. 

AEC 1958b, Symposium on Occupational Health Experience and Practices in the Uranium 

Industry, October 15-17, 1958. 

Anderson, J.L., and Hertel, N.E., “Bremsstrahlung Doses from Natural Uranium Ingots,” 

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Andersson K.G., et al., Radiation Dose Implications of Airborne Contaminant Deposition to 

Humans, Health Physics 82(2), pages 226-232; 2002. 

Barr, J.A., Jr., Ruch, J.W., and Borlik, R.F., “Recovering Uranium As By-Product in Phosphate 

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Blockson (Blockson Chemical Company) 1951, letter from E. B. Lopker, Director of 

Engineering, to Sheldon Wimpfen, Atomic Energy Commission, Raw Materials Operations, July 

31, 1951 [SRDB Ref ID 9558]. 
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Blockson 1953a, “Uranium Recovery from Wet Process Phosphoric Acid, Abstract,” prepared by 

Stoltz, Emil Jr., February 23, 1953 [SRDB Ref ID 4027]. 

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Boothe, Gary F., “The Need for Radiation Controls in the Phosphate and Related Industries,” 

Health Physics, Vol. 32, pp. 285-290, April 1977. 

Clegg, J.W., and Foley, D.D., Uranium Ore Processing, prepared under contract with the U.S. 

AEC, Addison-Wesley Publishing Company, Inc., 1958, pp. 375-378. 

Cope, David F., Acting Chief, Research Branch, and Sinclair, Edward E., Research Branch, U.S. 

AEC, record of “Conference of Uranium Recovery from Phosphate Rock, Tampa, Florida, 

December 8, 9, and 19, 1952,” December 30, 1952. 

Coppinger, E.A. and C.A. Rohrmann, Ionium (Thorium-230) for Radioisotope Preparation 

(Status Report), Publication HW-63239, 1959. 

DOE (U. S. Department of Energy) 1983, FUSRAP, “Radiological Survey of Chemicals Group, 

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28, 1978,” Argonne National Laboratory; DOE/EV-0005/35; ANL-OHS/HP-83-103, May 1983 

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in Uranium Facilities,” DOE-STD-1136-2000, Table 2-7, Washington, D. C., August 2000. 

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(DOE-STD-1136-2000, Change Notice No. 3, December 2001), 

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DOL (U. S. Department of Labor) 2007, Peter M. Turcic, Director, Division of Energy 

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EPA (U. S. Environmental Protection Agency) 1978, Radiological Surveys of Idaho Phosphate 

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ID 30886]. 
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EPA 1993a, Region 5, Air and Radiation Division, inspection report for 40 CFR 61, Subpart R 

compliance, September 7 [SRBD Ref ID 30131]. 

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FIPR (Florida Institute of Phosphate Research), 1995, Microbiology and Radiochemistry of 

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Fogh, Quantitative Measurement of Aerosol Deposition on Skin, Hair and Clothing for 

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Guimond, R. J. and S.T. Windham, “Radioactivity Distribution in Phosphate Products, Byproducts, Effluents, and Wastes, U.S. Environmental Protection Agency Technical Note 

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ICRP 1994a, “Human Respiratory Tract Model for Radiological Protection,” ICRP Publication 

66. 

ICRP 1994b, “Dose Coefficients for Intakes of Radionuclides by Workers,” ICRP Publication 

68. 

ICRP 1996, “Conversion Coefficients for Use in Radiological Protection Against External 

Radiation, ICRP Publication 74. 

Kocher, D.C., and Eckerman, K.F., “Electron Dose-Rate Conversion Factors for External 

Exposure of the Skin from Uniformly Deposited Activity on the Body Surface,” Health Physics, 

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NIOSH (National Institute for Occupational Safety and Health) 2006a, External Dose 

Reconstruction Implementation Guideline, OCAS-IG-001, Rev. 2, Office of Compensation 

Analysis and Support, Cincinnati, Ohio. 

NIOSH 2006b, “Lung Dose Conversion Factor for Thoron WLM,” OCAS-TIB-0011, Rev. 2, 

Office of Compensation Analysis and Support, Cincinnati, Ohio, January 13, 2006. 

NRC (U. S. Nuclear Regulatory Commission) 2002a, “Health Physics Surveys in Uranium 

Recovery Facilities,” Regulatory Guide 8.30, Revision 1, May 2002. 

NRC 2002b, “Technical Basis for Calculating Radiation Doses for the Building Occupancy 

Scenario Using the Probabilistic RESRAD-Build 3.0 Code, NUREG/CR-6755, February 2002. 

NRC 2002c, “Re-evaluation of the Indoor Resuspension Factor for the Screening Analysis of the 

Building Occupancy Scenario for NRC’s License Termination Rule, NREUG-1720, Draft Report 

for Comment, June 2002. 

OCAS (Office of Compensation Analysis and Support), OCAS-TIB-009, Technical Information 

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2007. 

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[Ref ID 31477]. 

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Ref ID 30132, p. 7]. 

Olin Corporation, letter to U.S. EPA with results of radon flux measurements conducted by 

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Florida Institute of Phosphate Research, 1984. 

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availability, Publication HW-66600, 1960. 

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